View source for LR-N16-0103, License Amendment Request to Revise Technical Specifications to Adopt TSTF-545, Revision 3, TS Inservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing Using Consolidated Line Item Improvement Process

{{Adams
| number = ML16203A006
| issue date = 07/20/2016
| title = License Amendment Request to Revise Technical Specifications to Adopt TSTF-545, Revision 3, TS Inservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing Using Consolidated Line Item Improvement Process
| author name = Davison P
| author affiliation = PSEG Nuclear, LLC
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000354
| license number = NPF-057
| contact person = 
| case reference number = LAR H16-05, LR-N16-0103
| document type = Letter type:LR, License-Application for Facility Operating License (Amend/Renewal) DKT 50
| page count = 20
| project = 
| stage = Request
}}

=Text=
{{#Wiki_filter:PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, NJ 08038-0236 PS~~G NudearLLC 10 CFR 50.90 LR-N16-0103 LAR H16-05
'JU( 202016 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Hope Creek Generating Station Renewed Facility Operating License No. NPF-57 NRC Docket No. 50-354

==Subject:==
License Amendment Request to Revise Technical Specifications to Adopt TSTF-545, Revision 3, "TS Inservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing" Using the Consolidated Line Item Improvement Process Pursuant to 10 CFR 50.90, PSEG Nuclear LLC (PSEG) is submitting a request for an amendment to the Technical Specifications (TS) to the renewed facility operating license listed above.
The proposed change revises the TS to eliminate TS 6.8.4.i, "Inservice Testing Program." A new defined term, "Inservice Testing Program (1ST)," is added to TS Definitions section. This request is consistent with TSTF-545, Revision 3, "TS Inservice Testing Program Removal &
Clarify SR (Surveillance Requirement) Usage Rule Application to Section 5.5 Testing."
By letter dated December 18, 2015, PSEG requested authorization to adopt American Society of Mechanical Engineers (ASME) Operation and Maintenance (OM) Code Case OMN-20, "Inservice Test Frequency," as an alternative to the ASME OM Code for the fourth 10 year Inservice Testing (1ST) interval which begins on December 21,2016. provides a description and assessment of the proposed TS changes. provides the existing TS pages marked up to show the proposed changes.
Approval of the proposed amendment is requested within one year of submittal. Once approved, the amendment shall be implemented within 60 days.
In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated State of New Jersey Official.

LR-N16-0103                                                                     10 CFR 50.90 Page 2 JUL 202016.
There are no regulatory commitments contained in this letter.
If you have any questions regarding this submittal, please contact Ms. Tanya Timberman at 856-339-1426.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on       J u \'1 (). D) J,.o "
{Date)
Respectfully, Paul Davison Site Vice President - Hope Creek Generating Station Attachments:
: 1. Description and Assessment of Technical Specification Changes
: 2. Proposed Technical Specification Changes (Mark-Up) cc:      Mr. D. Dorman, Administrator, Region I, NRC Ms. C. Parker, Project Manager, NRC NRC Senior Resident Inspector, Hope Creek Mr. P. Mulligan, Chief, NJBNE PSEG Corporate Commitment Tracking Coordinator Hope Creek Commitment Tracking Coordinator

LR-N16-0103                                                                                                          LAR H16-05 Attachment 1 Description and Assessment of Technical Specification Changes Table of Contents

==1.0    DESCRIPTION==
 ................................................................................................................1 2.0    ASSESSMENT ................................................................................................................1 2.1 Applicability of Published Safety Evaluation ............................................................. 1 2.2 Variations ................................................................................................................1

==3.0    REGULATORY ANALYSIS==
 .............................................................................................2 3.1  No Significant Hazards Consideration Analysis ....................................................... 2 4.0   ENVIRONMENTAL EVALUATION ................................................................................. .4

LR-N16-0103                                                                            LAR H16-05 

==1.0      DESCRIPTION==

The proposed change eliminates the Technical Specifications (TS), Section 6.8.4.i, "Inservice Test (1ST) Program," to remove requirements duplicated in American Society of Mechanical Engineers (ASME) Code for Operations and Maintenance of Nuclear Power Plants (OM Code),
Case OMN-20, "Inservice Test Frequency." A new defined term, "Inservice Testing Program,"
is added to TS Section 1.0, "Definitions". The proposed change to the TS is consistent with TSTF-545, Revision 3, "TS Inservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing."
2.0     ASSESSMENT 2.1     Applicability of Published Safety Evaluation PSEG has reviewed the model safety evaluation provided to the Technical Specifications Task Force in a letter dated December 11, 2015 (NRC ADAMS Accession No. ML15314A365). This review included a review of the NRC staff's evaluation, as well as the information provided in TSTF-545. PSEG concluded that the justifications presented in TSTF-545, and the model safety evaluation prepared by the NRC staff are applicable to Hope Creek Generating Station (Hope Creek) and justify this amendment for the incorporation of the changes to the Hope Creek TS.
Hope Creek was issued a construction permit, CPPR-120, on November 4, 1974 and the provisions of 10 CFR 50.55a(f)(3) are applicable.
2.2     Variations PSEG is not proposing any variations or deviations from the TS changes described in the TSTF-545 or the applicable parts of the NRC staff's model safety evaluation dated December 11,2015.
The Hope Creek TS utilize different numbering than the Standard Technical Specifications on which TSTF-545 was based. Specifically, the "Inservice Testing Program" in the Hope Creek TS is numbered 6.8.4.i rather than 5.5.7. The "Definitions" in the Hope Creek TS is numbered 1.0 rather than 1.1. Also, SR 3.0.2 and SR 3.0.3 are TS 4.0.2 and TS 4.0.3 in the Hope Creek TS.
As noted in Section 2.2.1 of the TSTF Traveler, the phrase "Inservice Testing Program" may appear in different locations in plant-specific TS. Revising this phrase to be capitalized wherever it may appear is within the scope of this proposed change. The SR listed in the Standard Technical Specifications differs from the SR listed in the Hope Creek TS as follows:
Standard Technical Specifications                  Hope Creek TS 3.1.7.7, SLC System                                4.1.5.c, Standby Liquid Control System 3.4.3.1, S/RVs 3.4.5.1, RCS PIV Leakage                           4.4.3.2.2, Operational Leakage 3.5.1.7, ECCS - Operating                          4.5.1.b, ECCS - Operating 3.5.2.5, ECCS - Shutdown 1

LR*N16*0103                                                                          LAR H16-05 Standard Technical Specifications                 Hope Creek TS 4.5.1.d.2.b, ECCS - Operating 3.6.1.3.6, PCIVs                                 4.6.3.3, Primary Containment Isolation Valves 3.6.1.3.8, PCIVs                                 4.4.7, Main Steam Isolation Valves 3.6.2.3.2, RHR Suppression Pool Cooling          4.6.2.3.b, Suppression Pool Cooling 3.6.2.4.2,  RHR Suppression Pool Spray            4.6.2.2.b, Suppression Pool Spray 3.6.4.2.2,  SCIVs 4.7.4.b, Reactor Core Isolation Cooling System Finally, TSTF-545 renumbers the Standard Technical Specifications sections to reflect elimination of the program and references to the renumbered are revised, whereas PSEG is keeping the existing numbering after deleting TS section 6.8.4.i.
These differences are administrative and do not affect the applicability of TSTF-545 to the Hope Creek TS.

==3.0      REGULATORY ANALYSIS==

3.1      No Significant Hazards Consideration PSEG requests adoption of the Technical Specification (TS) changes described in TSTF-545, "TS Inservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing," which is an approved change to the Improved Standard Technical Specifications (ISTS), into the Hope Creek TS. The proposed change revises the TS Chapter 6, "Administrative Controls," Section 6.8, "Procedures and Programs," to delete the "Inservice Testing (1ST) Program" specification. Requirements in the 1ST Program are removed, as they are duplicative of requirements in the American Society of Mechanical Engineers (ASME)
Operations and Maintenance (OM) Code, as clarified by Code Case OMN-20, "Inservice Test Frequency." Other requirements in Section 6.8 are eliminated because the Nuclear Regulatory Commission (NRC) has determined their appearance in the TS is contrary to regulations. A new defined term, "Inservice Testing Program," is added, which references the requirements of Title 10 of the Code of Federal Regulations (10 CFR), Part 50, paragraph 50.55a(f). PSEG has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:
: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No The proposed change revises TS Chapter 6, "Administrative Controls," Section 6.8, "Procedures and Programs," by eliminating the "Inservice Testing Program" specification.
Most requirements in the Inservice Testing Program are removed, as they are duplicative of requirements in the ASME OM Code, as clarified by Code Case OMN-20, "Inservice Test Frequency." The remaining requirements in the Section 6.8 1ST Program are eliminated because the NRC has determined their inclusion in the TS is contrary to regulations. A new defined term, "Inservice Testing Program," is added to the TS, which references the requirements of 10 CFR 50.55a(f).
2

LR-N16-0103                                                                              LAR 1-116-05 Performance of inservice testing is not an initiator to any accident previously evaluated. As a result, the probability of occurrence of an accident is not significantly affected by the proposed change. Inservice test frequencies under Code Case OMN-20 are equivalent to the current testing period allowed by the TS with the exception that testing frequencies greater than 2 years may be extended by up to 6 months to facilitate test scheduling and consideration of plant operating conditions that may not be suitable for performance of the required testing. The testing frequency extension will not affect the ability of the components to mitigate any accident previously evaluated as the components are required to be operable during the testing period extension. Performance of inservice tests utilizing the allowances in OMN-20 will not significantly affect the reliability of the tested components.
As a result, the availability of the affected components, as well as their ability to mitigate the consequences of accidents previously evaluated, is not affected.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
: 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No The proposed change does not alter the design or configuration of the plant. The proposed change does not involve a physical alteration of the plant; no new or different kind of equipment will be installed. The proposed change does not alter the types of inservice testing performed. In most cases, the frequency of inservice testing is unchanged.
However, the frequency of testing would not result in a new or different kind of accident from any previously evaluated since the testing methods are not altered.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
: 3. Do the proposed changes involve a significant reduction in a margin of safety?
Response: No The proposed change eliminates some requirements from the TS in lieu of requirements in the ASME Code, as modified by use of Code Case OMN-20. Compliance with the ASME Code is required by 10 CFR 50.55a. The proposed change also allows inservice tests with frequencies greater than 2 years to be extended by 6 months to facilitate test scheduling and consideration of plant operating conditions that may not be suitable for performance of the required testing. The testing frequency extension will not affect the ability of the components to respond to an accident as the components are required to be operable during the testing period extension. The proposed change will eliminate the existing TS 4.0.3 allowance to defer performance of missed inservice tests up to the duration of the specified testing frequency, and instead will require an assessment of the missed test on equipment operability. This assessment will consider the effect on a margin of safety (equipment operability). Should the component be inoperable, the TS provide actions to ensure that the margin of safety is protected. The proposed change also eliminates a 3

LR-N16-0103                                                                              LAR H16-05 Attachment 1 statement that nothing in the ASME Code should be construed to supersede the requirements of any TS. The NRC has determined that statement to be incorrect. However, elimination of the statement will have no effect on plant operation or safety.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
Based upon the above, PSEG concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.
4.0      ENVIRONMENTAL EVALUATION The proposed change would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change.
i i-4

LR-N16-0103                                                                     LAR H16-05 Attachment 2 Mark-up of Proposed Technical Specification Pages The following Technical Specifications pages for Renewed Facility Operating License NPF-57 are affected by this change request:
Technical Specification Index                                                           i 1.0, Definitions                                                1-3 4.1.5.c, Standby Liquid Control System                         3/4 1-20 4.4.3.2.2, Operational Leakage                                 3/4 4-12 4.4.7, Main Steam Isolation Valves                             3/4 4-26 4.5.1.b, ECCS - Operating                                      3/4 5-4 4.5.1.d.2.b, ECCS - Operating                                  3/4 5-5 4.6.2.2.b, Suppression Pool Spray                              3/4 6-15 4.6.2.3.b, Suppression Pool Cooling                            3/4 6-16 4.6.3.3, Primary Containment Isolation Valves                  3/4 6-18 4.7.4.b, Reactor Core Isolation Cooling System                 3/4 7-11 6.8.4.i, Inservice Testing Program                             6-16e

DEFINITIONS SECTION 1.0   DEFINITIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . PAGE 1.1   ACTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 1.2   DELETED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . " . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 1.3   AVERAGE PLANAR LINEAR HEAT GENERATION RATE . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 1.4   CHANNEL CALIBRATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 1.5   CHANNEL CHECK . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 1.6   CHANNEL FUNCTIONAL TEST . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 1.7   CORE ALTERATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-2 1.8   DELETED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1- 2 1.9   CORE OPERATING LIMITS REPORT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-2 1.10 CRITICAL POWER RATIO . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-2 1.11 DOSE EQUIVALENT 1-131 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-2 1.12 E-AVERAGE DISINTEGRATION ENERGy . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-2 1.13 EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME .................... 1-2 1.14 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME ............. 1-3 1.15 DELETED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3 1.16 DELETED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3 1.17 FREQUENCY NOTATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3 1.18 IDENTIFIED LEAKAGE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3 1.19 ISOLATION SYSTEM RESPONSE TIME . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3 1.20 LIMITING CONTROL ROD PATTERN . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3 1.21 LINEAR HEAT GENERATION RATE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4 1.22 LOGIC SYSTEM FUNCTIONAL TEST . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4 1 . 23 DELETED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1- 4 1.24 MEMBER (S) OF THE PUBLIC . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4 1.25 MINIMUM CRITICAL POWER RATIO . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4 1.18.1 INSERVICE TESTING PROGRAM ....................................................... 1-3 HOPE CREEK                                                     i                                   Amendment No.163

DEFrNITIONS EMERGENCY CORE COOLING SYSTEM (ECeS) RESPONSE TIME 1.13 The EMERGENCY CORE COOLING SYSTEN (EeeS) RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its Eees actuation setpoint at the channel senSor until the Eees equipment is capable of performing its safety function, i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc. Times shall includo diesel generator starting and sequence loading delays where app1:i.cable. The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured .
ElND-OF-CYCLE RECIRCULATION PUMP 'rRlP ~~YSTF.M RESPONSE TIME 1.14 The END-OF-CYCI.dl] RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME shall be that time interval to complete suppression of the electric arc between the fully open contacts of the recirculation pump circuit breaker from initial movement of the associated:
: a. Turbine stop valves, and
: b. Turbine control valves.
The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is
               'measured.
1.15 -DELETED i
1 . 16 ,DELETED                                                                     i 1
       '.FREQUENCY NOTATION
       .1.17 The FREQUENCY No'rATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1..
IDENTIFIED LEAKAGE 1 .18 IDENTIFIED LEAKAGE shall be:
: a. L,eakage into collection systems, such as pump seal or valve packing leaks, that is oaptured and conducted to a sump or collecting tank, or
: b. Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of the leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE.
ISOLATION SYSTEM RESPONSE TIME 1.19 The ISOLATION SYSTEM RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its isolation actuation setpoint at the channel sensor until the isolation valves travel to their required positions. Times shall include diesel generator starting and sequence loading delays where applicable. The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.
HOPE CREEK                                  1-3                   Amendment No.163 1.18.1 The INSERVICE TESTING PROGRAM is the licensee program that fulfills the re uirements of 10 CFR 50

REACTIVITY CONTROL SYSTEMS SURVEILLANCE RE.,QUIREMENTS (Continued)
: b. In accordance with the Surveillance Frequency Control Program by:
: 1.       Verifying the continuity of the explosive charge.
: 2.        Determining that the available weight of sodium penta borate is greater than or equal to 5,776 Ibs and the concentration of boron in solution is within the limits of Figure 3.1.5-1 by chemical analysis.*
: 3.        Verifying that each valve (manual, power operated or automatic) in the flow path that is not locked, sealed, or otherwise secured In position, is in its correct position.          PNSERVICE TESTING PROGRAM              I
: c. Demonstrating that, when tested pursuant to the 1ST Program, the minimum flow requirement of 41.2 gpm, per pump, at a pressure of greater than or equal to 1255 psig is met.
: d. In accordance with the Surveillance Frequency Control Program by:
: 1.       Initiating one of the standby liquid control system subsystem, including an explosive valve, and verifying that a flow path from the pumps to the reactor pressure vessel is available by pumping demineralized water into the reactor vessel and verifying that the relief valve does not actuate.
The replacement charge for the explosive valve shall be from the same manufactured batch as the one fired or from another batch which has been certified by having one of that batch successfully fired. Both injection subsystems shall be tested in accordance with the Surveillance Frequency Control Program.
: 2.       **Demonstrating that all heat traced piping between the storage tank and the injection pumps is unblocked and then draining and flushing the piping with demineralized water.
: 3.      Demonstrating that the storage tank heaters are OPERABLE by verifying the expected temperature rise of the sodium pentaborate solution in the storage tank after the heaters are energized.
This test shall also be performed anytime water or boron is added to the solution or when the solution temperature drops below 70&deg;F.
**   This test shall also be performed whenever both heat tracing circuits have been found to be inoperable and may be performed by any series of sequential, overlapping or total flow path steps such that the entire flow path is included.
HOPE CREEK                                                                    Amendment No. 187

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.3.2.1 The reactor coolant system leakage shall be demonstrated to be within each of the above limits by:
: a.      Monitoring the drywell atmospheric gaseous radioactivity in accordance with the Surveillance Frequency Control Program (not a means of quantifying leakage),
: b. Monitoring the drywell floor and eqUipment drain sump flow rate In accordance with the Surveillance Frequency Control Program, and
: c. Monitoring the drywell air coolers condensate flow rate in accordance with the Surveillance Frequency Control Program, and
: d. Monitoring the drywell pressure in accordance with the Surveillance Frequency Control Program (not a means of quantifying leakage), and
: e. Monitoring the reactor vessel head flange leak detection system in accordance with the Surveillance Frequency Control Program (not a means of quantifying leakage), and
: f. Monitoring the drywell temperature in accordance with the Surveillance Frequency Control Program (not a means of quantifying leakage).
4.4.3.2.2 Each reactor coolant system pressure isolation valve specified in Table 3.4.3.2"1 shall be demonstrated OPERABLE by leak testing pursuant to the 1ST Program and verifying the leakage of each valve to be within the specified limit: IINSERVICE TESTING PROGRAM                t
: a. In accordance with the Surveillance Frequency Control Program, and
: b. Prior to returning the valve to service following maintenance, repair or replacement work on the valve which could affect Its leakage rate.
The provisions of Specification 4.0.4 are not applicable for entry Into OPERATIONAL CONDITION 3.
4.4.3.2.3 The highllow pressure interface valve leakage pressure monitors shall be demonstrated OPERABLE with alarm setpoints per Table 3.4.3.2-2 by performance of a CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION at the frequencies specified in the Surveillance Frequency Control Program.
HOPE CREEK                                     3/44-12                          Amendment No. 193

REACTOR COOLANT SYSTEM 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.4.7 Two main steam line isolation valves (MSIVs) per main steam line shall be OPERABLE with closing times greater than or equal to 3 and less than or equal to 5 seconds.
APPLICABILITY: OPERATIONAL CONDITIONS 1,2 and 3.
ACTION:
: a.      With one or more MSIVs inoperable:
: 1. Maintain at least one M81V OPERABLE in each affected main steam line that is ppen and within 8 hours, either:
a)       Restore the inoperable valve(s) to OPERABLE status, or b)       Isolate the affected main steam line by use of a deactivated MSIV in the closed position.
: 2. Otherwise, be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
SURVEILLANCE REQUIREMENTS 4.4.7 Each of the above required MSIVs shall be demonstrated OPERABLE by verifying full closure between 3 and 5 seconds when tested pursuant to the ~
I r.::pN-:-::S=E=R:-:-""V:-:-::I C=E=-'=T'=E:::-=S=T::-:""I                                    A:-:'"M-:--11 N-:-:G:--:P=R::::"'"C::::"'"C=-=R::::"'"C OG HOPE CREEK                                      3/44-26                                                Amendment                         No. 185

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS 4.5,1 The emergency core cooling systems shall be demonstrated OPERABLE by:
: a.       In accordance with the Surveillance Frequency Control Program:
: 1.      For the core spray system, the LPCI system, and the HPCI system:
a)       Verifying by venting at the high point vents that the system piping from the pump discharge valve to the system isolation valve is filled with water, b)       Verifying that each valve, manual, power operated or automatic, in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct* position, c)       Verify the RHR System cross tie valves on the discharge side of the pumps are closed and power, if any, is removed from the valve operators.
2,       For the HPCI system, verifying that the HPCI pump flow controller is in the correct position.
: b.       Verifying that, when tested pursuant to the 1ST Program: IINSERVICE TESTING PROGRAM             I
: 1.      The two core spray system pumps in each subsystem together develop a flow of at least 6150 gpm against a test line pressure corresponding to a reactor vessel pressure of ~105 psi above suppression pool pressure.
: 2.       Each LPCI pump in each sUbsystem develops a flow of at least 10,000 gpm against a test line pressure corresponding to a reactor vessel to primary containment differential pressure of 220 psid.
: 3.       The HPCI pump develops a flow of at least 5600 gpm against a test line pressure corresponding to a reactor vessel pressure of 1000 psig when steam Is being supplied to the turbine at 1000, +20, -80 psig. *"
c,       In accordance with the Surveillance Frequency Control Program:
: 1.       For the core spray system, the LPCI system, and the HPCI system, performing a system functional test which includes simulated automatic actuation of the system throughout its emergency operating sequence and verifying that each automatic valve in the flow path actuates to its correct position. Actual injection of coolant into the reactor vessel may be excluded from this test.
* Except that an automatic valve capable of automatic return to its ECCS position when an ECCS signal is present may be in position for another mode of operation.
The provisions of Specification 4.0.4 are not applicable provided the surveillance is performed within 12 hours after reactor steam pressure is adequate to perform the test.
HOPE CREEK                                           3/4 5-4                           Amendment No. 187

EMERGENCY CORE COOLING SYSTEMS B..URVEILLANCE REQUI REMENT&sect;(Contlnued)
: 2.       For the HPCI system, verifying that:
a)      The system develops a flow of at least 5600 gpm against a test line pressure corresponding to a reactor vessel pressure of
:::200 pslg, when steam is being supplied to the turbine at 200 + 15, -0 psig.**
b)     The suction is automatically transferred from the condensate storage tank to the suppression chamber on a condensate storage tank water level - low signal and on a suppression chamber -
water level high signal.
: 3.       Performing a CHANNEL CALIBRATION of the CSS, and LPCI system discharge line "keep filled" alarm instrumentation.
: 4.      Performing a CHANNEL CALI BRATION of the CSS header L\P instrumentation and verifying the setpoint to be ~ the allowable value of 4.4 psid.
: 5.      Performing a CHANNEL CALIBRATION of the LPCI header L\P instrumentation and verifying the setpoint to be ~ the allowable value of 1.0 psid.
: d. For the ADS:
: 1.      In accordance with the Surveillance Frequency Control Program, performing a CHANNEL FUNCTIONAL TEST of the Primary Containment Instrument Gas System lOW-low pressure alarm system.
: 2.      In accordance with the SUrveillance Frequency Control Program:
a)      Performing a system functional test which includes simulated automatic actuation of the system throughout its emergency operating sequence, but excluding actual valve actuation.
b)       Verify that when tested pursuant to the 1ST Program, that each ADS valve is capable of being opened. jlNSERVICE TESTING PROGRAM  I c)       Performing a CHANNEL CALIBRATION of the Primary Containment Instrument Gas System lOW-low pressure alarm system and verifying an alarm setpoint of 85 +/- 2 psig on decreasing pressure.
**     The provisions of Specification 4.0.4 are not applicable provided the surveillance is performed within 12 hours after reactor steam pressure is adequate to perform the test.
HOPE CREEK                                                                 Amendment No. 187

CONTAINMENT SYSTEMS SUPPRESSION POOL SPRAY LIMITING CONDITION FOR OPERATION 3.6.2.2 The suppression pool spray mode of the residual heat removal (RHR) system shall be OPERABLE with two independent loops, each loop consisting of:
: a.      One OPERABLE RHR pump, and
: b.      An OPERABLE flow path capable of recirculating water from the suppression chamber through an RHR heat exchanger and the suppression pool spray sparger.
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3.
ACTION:
: a.      With one suppression pool spray loop inoperable, restore the inoperable loop to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
: b.      With both suppression pool spray loops inoperable, restore at least one loop to OPERABLE status within 8 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN* within the following 24 hours.
SURVEILLANCE REQUIREMENTS 4.6.2.2 The suppression pool spray mode of the RHR system shall be demonstrated OPERABLE:
: a.      In accordance with the Surveillance Frequency Control Program by verifying that each valve, manual, power operated or automatic, in the flow path that is not locked, sealed or otherwise secured in position, is in its correct position.
: b.      By verifying that each of the required RHR pumps develops a flow of at least 540 gpm on recirculation flow through the RHR heat exchanger (after consideration of flow through the closed bypass valve) and suppression pool spray sparger when tested pursuant to the 1ST ProgFaffl.
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* Whenever both RHR subsystems are inoperable, if unable to attain COLD SHUTDOWN as required by this ACTION, maintain reactor coolant temperature as low as practical by use of alternate heat removal methods.
HOPE CREEK                                    3/4 6-15                                             Amendment No. 187

CONTAINMENT SYSTEMS SUPPRESSION POOL COOLING LIMITING CONDITION FOR OPERATION 3.6.2.3 The suppression pool cooling mode of the residual heat removal (RHR) system shall be OPERABLE with two independent loops, each loop consisting of:
: a.       One OPERABLE RHR pump, and
: b.       An OPERABLE flow path capable of recirculating water from the suppression chamber through an RHR heat exchanger.
APPLICABILITY: OPERf.TIONAL CONDITIONS 1, 2 and 3.
ACTION:
: a.      With one suppression pool cooling loop inoperable, restore the inoperable loop to OPERABLE status within 72 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours.
: b.      With both suppression pool cooling loops inoperable, be in at least HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN* within the next 24 hours.
SURVEILLANCE REQUIREMENTS 4.6.2.3 The suppression pool cooling mode of the RHR system shall be demonstrated OPERABLE:
: a.      In accordance with the Surveillance Frequency Control Program by verifying that each valve, manual, power operated or automatic, in the flow path that is not locked, sealed or otherwise secured in position, is in its correct position.
: b.      By verifying that each of the required RHR pumps develops a flow of at least 10,160 gpm on recirculation flow through the RHR heat exchanger (after consideration of flow through the closed bypass valve) and the suppression pool when tested pursuant to the 1ST Pro@faffi .
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* Whenever both RHR subsystems are inoperable, if unable to attain COLD SHUTDOWN as required by this ACTION, maintain reactor coolant temperature as low as practical by use of alternate heat removal methods.
HOPE CREEK                                                                                                                   Amendment No. 187

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS 4.6.3.1 Each primary containment isolation valve shall be demonstrated OPERABLE prior to returning the valve to service after maintenance, repair or replacement work is performed on the valve or its associated actuator, control or power circuit by cycling the valve through at least one complete cycle of full travel and verifying the specified isolation time.
4.6.3.2 Each primary containment automatic isolation valve shall be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program by verifying that on a containment isolation test signal each automatic isolation valve actuates to its isolation position.
4.6.3.3 The isolation time of each primary containment power operated or automatic valve shall be determined to be within its limit when tested pursuant to the ~~~~ra~m~.~-=-=::-::-::=-=""c-::-=~~"""""""",
INSERVICE TESTING PROGRAM 4.6.3.4 In accordance with the Surveillance Frequency Control Program, veri y a a representative sample of reactor instrumentation line excess flow check valves# actuates to the isolation position on a simulated instrument line break signal.
4.6.3.5 Each traversing in-core probe system explosive isolation valve shall be demonstrated OPERABLE*:
: a.      In accordance with the Surveillance Frequency Control Program by verifying the continuity of the explosive charge.
: b.      In accordance with the Surveillance Frequency Control Program by removing the explosive squib from at least one explosive valve, and initiating the explosive squib. The replacement charge for the exploded squib shall be from the same manufactured batch as the one fired or from another batch which has been certified by having at least one of that batch successfully fired. No squib shall remain in use beyond the expiration of its shelf-life or operating life, as applicable.
* Exemption to Appendix J of 10 CFR Part 50.
#       The reactor vessel head seal leak detection line (penetration J5C) is not required to be tested pursuant to this requirement.
HOPE CREEK                                      3/46-18                          Amendment No. 187

PLANT SYSTEMS 3/4.7.4 REACTOR CORE ISOLATION COOLING SYSTEM LIMITING CONDITION FOR OPERATION 3.7.4 The reactor core isolation cooling (RCIC) system shall be OPERABLE with an OPERABLE flow path capable of automatically taking suction from the suppression pool and transferring the water to the reactor pressure vessel.
APPLICABILITY: OPERATIONAL CONDITIONS 1,2, and 3 with reactor steam dome pressure greater than 150 psig.
ACTION:
Note: LCO 3.0.4.b is not applicable to RCIC.
With the RCIC system inoperable, operation may continue provided the HPCI system is OPERABLE; restore the RCIC system to OPERABLE status within 14 days or be in at least HOT SHUTDOWN within the next 12 hours and reduce reactor steam dome pressure to less than or equal to 150 psig within the following 24 hours.
SURVEILLANCE REQUIREMENTS 4.7.4 The RCIC system shall be demonstrated OPERABLE:
: a.       In accordance with the Surveillance Frequency Control Program by:
: 1. Verifying by venting at the high point vents that the system piping from the pump discharge valve to the system isolation valve is filled with water.
: 2.      Verifying that each valve, manual, power operated or automatic in the flow path that is not locked, sealed or otherwise secured in position, is in its correct position.
: 3.      Verifying that the Rum flow controller is in the correct osition.
INSERVICE TESTING PROGRAM
: b.      When tested pursuant to the ~ by verifying that the RCIC pump develops a flow of greater than or equal to 600 gpm in the test flow path with a system head corresponding to reactor vessel operating pressure when steam is being supplied to the turbine at 1000 + 20, - 80 psig.*
* The provisions of Specification 4.0.4 are not applicable provided the surveillance is performed within 12 hours after reactor steam pressure is adequate to perform the test.
HOPE CREEK                                      3/47-11                        Amendment No. 187

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued) 6.8.4.i     INSERVICE TESTING PROGRAM ~Deleted
::rhis Program provides controls for inse-Priee testing of ASME Code Class 1, 2, aRa 3 oompononts. ::rhe program shall incluae tho following:
a:        ::resting frequenoies applioable to the ASME Code for Operations and Maintenanoe of ~JuGlear Powor Plants (i\SME OM Code) and applioablo Addenda as follo'#s:
ASME OM Code aRd applioablo            Required Frequencies for Addenda terminology for                performing inservioe inservice testing aotivities           testing activities Weel<;ly                               At least enoe per 7 days M~                                     At least once per a1 days Quarterly or every 2 mORths            At-ieast onco per 92 days SemianRually or every 6 months         At least ono~r 184 days every 9 mORths                         At-Ieast onoe-per 276 days
                        &#xa5;early or annually                     At least onee FIeF a66 days Biennially or every 2 years            At-Ieast onee FIeF 7d1 days b-:      ::rhe provisions of Speoifioation 4.0.2 are applioable to tho above required frequencies and to other normal and acoeler~ted frequenoies specifiea as 2 years or less iR the IRservice ::restiRg Program for performing iRservice testing aativities, 1,
::rho provisions of Speoifioation 4.0.a aro aPFllioable to inservioe testing ootivities, and l-:      Nothing in the ASME OM Code shall be construed to supersede the requirements of any Teohnioal SpeolfioatiGFh 6.8.4.j    Surveillance Frequency Control Program This program provides controls for Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met.
: a.       The Surveillance Frequency Control Program shall contain a list of Frequencies of those Surveiilance Requirements for which the Frequency is controlled by the program.
: b.       Changes to the Frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI 04"10, "Risk-Informed Method for Control of Surveillance Frequencies," Revision 1.
: c.       The provisions of Surveillance Requirements 4.0.2 and 4.0.3 are applicable to the Frequencies established in the Surveillance Frequency Control Program.
HOPE CREEK                                     6"16e                            Amendment No. 187}}